The tokamak approach, utilizing a toroidal magnetic field configuration to confine a hot plasma, is one of the most promising designs for developing reactors that can exploit nuclear fusion to generate electrical energy(1,2). To reach the goal of an economical reactor, most tokamak reactor designs(3-10) simultaneously require reaching a plasma line-averaged density above an empirical limit-the so-called Greenwald density(11)-and attaining an energy confinement quality better than the standard high-confinement mode(12,13). However, such an operating regime has never been verified in experiments. In addition, a long-standing challenge in the high-confinement mode has been the compatibility between a high-performance core and avoiding large, transient edge perturbations that can cause very high heat loads on the plasma-facing-components in tokamaks. Here we report the demonstration of stable tokamak plasmas with a line-averaged density approximately 20% above the Greenwald density and an energy confinement quality of approximately 50% better than the standard high-confinement mode, which was realized by taking advantage of the enhanced suppression of turbulent transport granted by high density-gradients in the high-poloidal-beta scenario(14,15). Furthermore, our experimental results show an integration of very low edge transient perturbations with the high normalized density and confinement core. The operating regime we report supports some critical requirements in many fusion reactor designs all over the world and opens a potential avenue to an operating point for producing economically attractive fusion energy.

Internal fluctuation measurements with Faraday-effect polarimetry in the DIII-D tokamak reveal the onset of a tearing mode with toroidal mode number n = 3 well before it is detected by the sensing coils external to the plasma. This mode appears before the n = 2, 1 modes and is first detected with internal measurements at a lower value of the ideal-wall kink beta limit than is indicated at the time of first detection by the sensing coils. When the mode is first detected, the linear resistive stability parameter, Δ′, indicates marginal stability and continues to do so until later when the mode amplitude begins increasing linearly with time—together suggesting a neoclassical origin for this mode.

Since the last IAEA-FEC in 2021, significant progress on the development of long pulse steady state scenario and its related key physics and technologies have been achieved, including the reproducible 403 s long-pulse steady-state H-mode plasma with pure radio frequency (RF) power heating. A thousand-second time scale (similar to 1056 s) fully non-inductive plasma with high injected energy up to 1.73 GJ has also been achieved. The EAST operational regime of high beta(P) has been significantly extended (H-98y2 > 1.3, beta(P) similar to 4.0, beta(N) similar to 2.4 and n(e)/n(GW) similar to 1.0) using RF and neutral beam injection (NBI). The full edge localized mode suppression using the n = 4 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with q(95) approximate to 3.1 on EAST, extrapolating favorably to the ITER baseline scenario. The sustained large ELM control and stable partial detachment have been achieved with Ne seeding. The underlying physics of plasma-beta effect for error field penetration, where toroidal effect dominates, is disclosed by comparing the results in cylindrical theory and MARS-Q simulation in EAST. Breakdown and plasma initiation at low toroidal electric fields (<0.3 V m(-1)) with EC pre-ionization is developed. A beneficial role on the lower hybrid wave injection to control the tungsten concentration in the NBI discharge is observed for the first time in EAST suggesting a potential way toward steady-state H-mode NBI operation.

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant beta phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent 'pedestal-to-divertor' integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T-e and radiation which need control. High-beta(P) scenarios with q(min) > 2, q(95)-7.9, beta(N) > 4, beta(T)-3.3% and H-98y2 > 1.5 are sustained with high density ((n) over bar = 7E19 m(-3), f(G)-1) for 6 tau(E), improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international C oordination on I nternational C hallenges on L ong duration OP eration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.

To address the needs for a fusion pilot plan design, DIII-D/EAST joint experiments on DIII-D have demonstrated high normalized beta β N ∼ 4.2, toroidal beta β T ∼ 3.3% with q min > 2, q 95 ⩽ 8 sustained for more than six energy confinement times in high poloidal beta regime. The excellent energy confinement quality ( H 98y2 ∼ 1.8) is achieved with an internal transport barrier at high line-averaged Greenwald density fraction f Gr > 0.9. The trapped gyro-Landau fluid (TGLF) modeling of the transport characteristics shows that the beam-driven rotation does not play an important role in the high confinement quality. The modeling also captures very well several transport features, giving us confidence in using integrated modeling to project these experimental results to future machines. The high-performance phase is terminated by fast-growing modes triggered near the n = 1 ideal-wall kink stability limit. New radio frequency (RF) capabilities for off-axis current drive could remove the residual ohmic current to achieve a fully non-inductive state, and improve the mode–wall coupling to increase the ideal-wall β N limit, enabling sustainment of the fully non-inductive high performance plasma in stationary conditions.

Analyses of the DIII-D ITER Baseline Scenario database support that the disruptive m,n =2,1 magnetic islands are pressure gradient driven, non-linear instabilities seeded in a sequence of stochastic transient magnetic perturbations, and that the current profile relaxation does not affect the m,n =2,1 island onset rate. At low torque, these Neoclassical Tearing Modes are most commonly seeded by non-linear 3-wave coupling when the differential rotation between the q =1 & q =2 rational surfaces approaches zero. Lack of statistically significant difference between the current profiles of stable and unstable states, as well as lack of correlation between the tearing mode onset rate and the current profile relaxation both reject causality between the current profile evolution and the 2,1 magnetic island onsets in these plasmas. These support that preserving the differential rotation between the q =1 and q =2 rational surfaces is key to long pulse stable operation in the plasma scenario planned for ITER, while optimization of the current profile within the explored parameter space may lead to much weaker improvements than sustaining the differential rotation.

A record duration of a 310 s H-mode plasma (H98y2 ∼ 1.3, ne/nGW ∼ 0.7, fBS > 50%) has been recently achieved on experimental advanced superconducting tokamak (EAST) with metal walls, exploiting the device's improved long-pulse capabilities. The experiment demonstrates good control of tungsten concentration, core/edge MHD stability, and particle and heat exhaust with an ITER-like tungsten divertor and zero injected torque, establishing a milestone on the path to steady-state long-pulse high-performance scenarios in support of ITER and CFETR. Important synergistic effects are leveraged toward this result, which relies purely on radio frequency (RF) powers for heating and current drive (H&CD). On-axis electron cyclotron heating enhances the H&CD efficiency from lower hybrid wave injection, increasing confinement quality and enabling fully non-inductive operation at high density (ne/nGW ∼ 70%) and high poloidal beta (βP ∼ 2.5). A small-amplitude grassy edge localized mode regime facilitates the RF power coupling to the H-mode edge and reduces divertor sputtering/erosion. The high energy confinement quality (H98y2 ∼ 1.3) is achieved with the experimental and simulated results pointing to the strong effect of Shafranov shift on turbulence. Transport analysis suggests that trapped electron modes dominate in the core region during the record discharge. The detailed physics processes (RF synergy, core-edge integration, confinement properties, etc.) of the steady-state operation will be illustrated in the content. In the future, EAST will aim at accessing more relevant dimensionless parameters to develop long-pulse high-performance plasma toward ITER and CFETR steady-state advanced operation.

The path to fusion in the United States requires partnership between public and private sector. While the private sector provides the vigor to take some of the major steps necessary, there is a depth of expertise and capability in the public sector that is vital to resolving feasible approaches. As an open national user facility, DIII-D provides a crucial testbed to develop the required new technologies and approaches in relevant conditions. It has unparalleled potential to meet this challenge, thanks to its extreme flexibility and world leading diagnostics. This provides a basis to rapidly develop solutions that project to future reactors with confidence. The program has thus been redeveloped to enable public and private sector engagement and testing of new concepts. A new technology program has been launched to resolve plasma interacting technologies. With modest heating upgrades, the facility can confront the crucial "Integrated Tokamak Exhaust and Performance" gap, to resolve core, exhaust and technology solutions together. The device is also being redeveloped as a training facility, with dedicated student run time, a mentorship program, and open access to all opportunity roles, part of wider efforts to diversify and open pathways through inclusion, access, and equity. This exciting agenda is enabling scientists and technology researchers to pioneer the solutions needed for a Fusion Pilot Plant (FPP) and ITER this decade. As a national user facility, DIII D has singular potential to provide the tools, teams, and insight necessary, to do its part in moving the United States rapidly toward the commercialization of fusion energy.

A recent EAST experiment has successfully demonstrated long pulse steady-state high plasma performance scenario and core-edge integration since the last IAEA in 2018. A discharge with a duration over 60 s with β P ∼ 2.0, β N ∼ 1.6, H 98y2 ∼ 1.3 and an internal transport barrier on the electron temperature channel is obtained with multi-RF power heating and current drive. A higher β N ( β N ∼ 1.8, β p ∼ 2.0, H 98y2 ∼ 1.3, n e / n GW ∼ 0.75) with a duration of 20 s is achieved by using the modulated neutral beam and multi-RF power, where several normalized parameters are close or even higher than the phase III 1 GW scenario of CFETR steady-state. High-Z impurity accumulation in the plasma core is well controlled in a low level by using the on-axis ECH. Modeling shows that the strong diffusion of TEM turbulence in the central region prevents tungsten impurity from accumulating. More recently, EAST has demonstrated compatible core-edge integration discharges in the high β p scenario: high confinement H 98y2 > 1.2 with high β P ∼ 2.5/ β N ∼ 2.0 and f bs ∼ 50% is sustained with reduced divertor heat flux at high density n e / n GW ∼ 0.7 and moderate q 95 ∼ 6.7. By combining active impurity seeding through radiative divertor feedback control and strike point splitting induced by resonant perturbation coil, the peak heat flux is reduced by 20–30% on the ITER-like tungsten divertor, here a mixture of 50% neon and 50% D 2 is applied.

DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter- I p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high- Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L–H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ∼ 8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co- I p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.

For the first time, experiments on the DIII-D tokamak have demonstrated electron cyclotron current drive with more than double the conventional efficiency by tailoring the wave–particle interactions in velocity space using a novel ‘top launch’ geometry. Steering the EC waves to propagate nearly parallel to the resonance drives current more efficiently by (1) selective damping on electrons with higher parallel velocity v ||, and (2) longer absorption path to compensate for inherently weaker absorption at higher v ||. Experiments using a fixed-injection top launch system find an optimal velocity space interaction for maximum current drive efficiency at ρ ∼ 0.5 where the ease of drawing out a high v || electron tail is balanced by sufficient absorption.

A new diagnostic technique has been developed that couples a spectrometer and an image-intensified camera into the traditional motional Stark effect (MSE) system on DIII-D. The image-intensified camera syncs with the photo-elastic modulators to spectrally resolve the Stokes parameters across the Stark multiplet. Polarization dependent phase shift, likely from a plasma facing mirror, leads to the spectropolarimeter measuring a variation in the polarization angle across the MSE spectrum of ∼8°.

Recent Progress in High Poloidal Beta Scenario Development on DIII-D S. Ding1, A. M. Garofalo2, X. Gong3, J. Qian3, J. Huang3, H. Wang2, L. Wang3, C. T. Holcomb4, X. Jian5, C. Pan3, Q. Ren3, G. McKee6, D. Weisberg2, D. Eldon2 1Oak Ridge Associated University, 2General Atomics, 3Institute of Plasma Physics, Chinese Academy of Sciences, 4Lawrence Livermore National Laboratory, 5University of California, San Diego, 6University of Wisconsin-Madison e-mail (speaker): dingsiye@fusion.gat.com

Physics-based simulations project a compact net electric fusion pilot plant with a nuclear testing mission is possible at modest scale based on the advanced tokamak concept, and identify key parameters for its optimization. These utilize a new integrated 1.5D core-edge approach for whole device modeling to predict performance by self-consistently applying transport, pedestal and current drive models to converge fully non-inductive stationary solutions, predicting profiles and energy confinement for a given density. This physics-based approach leads to new insights and understanding of reactor optimization. In particular, the levering role of high plasma density is identified, which raises fusion performance and self-driven 'bootstrap currents', to reduce current drive demands and enable high pressure with net electricity at a compact scale. Solutions at 6-7 T, similar to 4 m radius and 200 MW net electricity are identified with margins and trade-offs possible between parameters. Current drive comes from neutral beam and ultra-high harmonic (helicon) fast wave, though other advanced approaches are not ruled out. The resulting low recirculating power in a double null configuration leads to a divertor heat flux challenge that is comparable to ITER, though reactor solutions may require more dissipation. Strong H-mode access (x2 margin over L-H transition scalings) and ITER-like heat fluxes are maintained with similar to 20%-60% core radiation, though effects on confinement need further analysis. Neutron wall loadings appear tolerable. The approach would benefit from high temperature superconductors, as higher fields would increase performance margins while potential for demountability may facilitate nuclear testing. However, solutions are possible with conventional superconductors. An advanced load sharing and reactive bucking approach in the device centerpost region provides improved mechanical stress handling. The prospect of an affordable test device which could close the loop on net-electric production and conduct essential nuclear materials and breeding research is compelling, motivating research to validate the techniques and models employed here.

A slowly rotating 3D field has been considered a promising approach for preventing the locking and growth of NTMdriven disruptions in high beta plasmas, or enhancing the H-mode recovery. Recently, M3D [1], numerical simulations have found new properties of the q=1 modes in these plasmas that suggest the possibility of using the rotating 3D field to sustain the plasma core against quasi-interchange (QI)-driven sawtooth crashes, at the same time as avoiding NTMs locking at q>1. The characteristic q~1 QI modes are found to have dominant n=1 and 2 toroidal harmonics of comparable magnitude that are well suited to interact with the external 3D field. Preliminary experimental observations in hybrid configuration discharges support the possibility of quasi-interchange mode control with the slowly rotating 3D field and suggest that it can be explored simultaneously with the control of NTM locking avoidance.

To prepare for steady-state operation of future fusion reactors (e.g. the International Thermonuclear Experimental Reactor and China Fusion Engineering Test Reactor (CFETR)), experiments on DIII-D have extended the high poloidal beta (beta(P)) scenario to reactor-relevant edge safety factorq(95)similar to 6.0, while maintaining a large-radius internal transport barrier (ITB) using negative magnetic shear. Excellent energy confinement quality (H-98y2> 1.5) is sustained at high normalized beta (beta(N)similar to 3.5). This high-performance ITB state with Greenwald density fraction near 100% andq(min)>= 3 is achieved with toroidal plasma rotationV(tor)similar to 0 at rho >= 0.6. This is a key result for reactors expected to have lowV(tor). At high beta(P)(>= 1.9), large Shafranov shift can stabilize turbulence leading to a high confinement state with a low pedestal and an ITB. At lower beta(P)(<1.9), negative magnetic shear in the plasma core contributes to turbulence suppression and can compensate for reduced Shafranov shift to continue to access a large-radius ITB and excellent confinement with lowV(tor), consistent with the results of gyrofluid transport simulations. These high-beta(P)cases are characterized by weak/no Alfven eigenmodes (a.e.) and classical fast-ion transport. At high density, the fast-ion deceleration time decreases and Delta beta(fast)is lower; these reduce a.e. drive. The reverse-shear Alfven eigenmodes are weaker or stable because the negative magnetic shear region is located at higher radius, away from the peaked fast-ion profile. Resistive wall modes can be a limitation at simultaneous high beta(N), low internal inductance, and low rotation. Analysis suggests that additional off-axis external current drive could provide a more stable path at reducedq(95). Based on a DIII-D high-beta(P)plasma with large-radius ITB, two scenarios are proposed for CFETRQ= 5 steady-state operation with similar to 1 GW fusion power: a lower-li <i(li<i similar to 0.66) and a higher-li<i(li<i similar to 0.75) case. Using a Landau closure model, multiple energetic particle (EP) effects on the a.e. stability are analyzed modifying the growth rate of the a.e.s triggered by the neutral-beam-injection EPs and alpha particles, although the stabilizing/destabilizing effect is weak for the cases analyzed. The stabilizing effects of the combined EP species beta, energy, and density profile in CFETR need further investigation.